APR+ 2013 06 10 (final from Mr. Seo) - ARIS

39 Pages·2013·1.04 MB·English

APR + (Advanced Power Reactor Plus)

Korea Hydro and Nuclear Power Company, Republic of Korea

Overview

Full name Advanced Power Reactor Plus
Acronym APR +
Reactor type PWR
Coolant Light Water
Moderator Light Water
Neutron Spectrum Thermal Neutrons
Thermal Capacity 4290 MW(th)
Electrical Capacity 1560 MW(e) gross
1505 MW(e) net
Design Organization Korea Hydro and Nuclear Power Co.
Last update 11-06-2013

1. Introduction

The Advanced Power Reactor Plus (APR+), an improved nuclear power reactor with 1500MW
electric power to succeed the Republic of Korea’s current Advanced Power Reactor 1400
(APR1400), has been developed as a two-loop evolutionary pressurized water reactor by adopting a
number of advanced design features to further enhance safety, economics, and reliability.
The Korean government and nuclear industry cooperatively launched the “Nuclear
Technology Development Plan 2012 (Nu-Tech 2012 plan)” in December 2006. The goal of
this plan is to enhance technical competitiveness in the global market and to have an
intellectual property-free reactor design. The main project under the Nu-Tech 2012 plan is the
development of APR+ standard design and related key technologies. The APR+ technology
development project consists of three sub-projects: 1) the feasibility study for the APR+
development, 2) the key technology development for the APR+ design, and 3) the APR+ standard
design development.
The first sub-project, feasibility study for APR+ development, has been carried out for 2 years from
August 2007 to July 2009. The major goals of this study are to identify top-tier requirements, to
develop a conceptual design, and to evaluate safety, economics, and performance characteristics of
the APR+ conceptual design. From this study, it is concluded that the APR+ design can be
developed as a two loop evolutionary pressurized water reactor with a number of advanced design
features to enhance safety and economics based on the APR1400 technology. For the economic
enhancement, the APR+ reactor core power has been determined to be increased to 4,290 MWth
which corresponds to a 1500MWe class nuclear power plant. Also, the plant design development has
been performed so that several new construction technologies can be introduced in order to shorten
the construction period.
For the safety improvement, a number of advanced design features are introduced in the APR+
design such as the improved direct vessel injection (DVI+), advanced fluidic device (FD+), passive
auxiliary feedwater system (PAFS), and four independent trains of safety systems, mechanically as
well as electrically, based on N+2 design concept. As severe accident mitigation design features, the
emergency reactor depressurization system (ERDS) for rapid depressurization during high pressure
severe accident scenarios and the enhanced in-vessel retention through external reactor vessel
cooling (IVR-ERVC) system are newly incorporated. Also, technical countermeasures against the
unexpected beyond design basis natural disaster have been selected for incorporation into the APR+
standard design. The major design requirements for the safety and performance goals set for APR+
are listed in Table 1.
TABLE 1. APR+ DESIGN REQUIREMENTS FOR SAFETY AND PERFORMANCE GOALS

General Requirement Performance requirements and economic goals
Type and capacity : PWR, 1500 MWe Plant availability : greater than 92%
Plant lifetime : 60 years Unplanned trips : less than 0.2 times per year
Seismic design : SSE 0.3g Refueling interval : 18 months or longer
Safety goals : Construction period : 36 months (N-th plant) from
Core damage frequency < 1.0E-6/RY first concrete (F/C) to fuel loading (F/L)
Containment failure frequency < 1.0E-7/RY
Economic goal : ≥ 20% cost advantage over
Occup. radiation exposure < 1 man -Sv / RY
fossil fueled power plants

The second sub-project, key technology development for the APR+ design, is intended to develop
and demonstrate major advanced design features in order to support APR+ standard design
development focusing on six (6) technologies: 1) safety injection system optimization and
development including the improved Direct Vessel Injection (DVI+) and the advanced fluidic device
(FD+), 2) passive auxiliary feedwater system (PAFS), 3) multiple severe accident mitigation system,
4) automatic load follow operation, 5) improved control element drive mechanism (CEDM), and 6)
combined modularization of components and structure. Finally, the third sub-project, APR+ standard
design development, launched in April 2009 aiming at acquiring the standard design approval from
the Korean nuclear regulatory body by 2013.
The APR+ RCS configuration is the same as that of the APR1400 as shown in Figure 1. The APR+
standard design development is completed by incorporating the key technologies developed from the
second sub-project, the state-of-the-art design features accepted globally and the design
improvements and lessons learned from design and construction of the APR1400. The first
APR1400 units are being built in the Republic of Korea as Shin-Kori Nuclear Power units 3 and 4
(SKN 3&4). Additionally the self-developed new reactor coolant pump (RCP) and man machine
interface system (MMIS) technologies will be implemented in the APR+ design. Also, the reactor
core design and safety analyses are performed using the new specific code packages which are also
under development in parallel with the APR+ development. Finally, the increased reactor power up
to 4,290 MWth will be accomplished by adding sixteen (16) fuel assemblies to the APR1400 reactor
core.
Figure 1. APR+ Reactor Coolant System Configuration

2. Description of the Nuclear Systems

2.1 Primary circuit and its main characteristics
The APR+ is characterized as a two-loop evolutionary PWR. The APR+ core thermal power
is uprated to 4,290 MWth, which corresponds to a 1,500 MWe class nuclear power plant. This
power rating is 108% of the APR1400 core power and is considered to be the maximum
power output with a two-loop reactor coolant system (RCS) configuration while minimizing
component size change. As shown in Figure 1, the APR+ RCS contains two primary coolant
loops, each of which consists of one 1.07-meter (42-inch) inside diameter (ID) hot leg, two
0.76-meter (30-inch) ID cold legs, one steam-generator (SG) and two reactor coolant pumps
(RCPs). One pressurizer (PZR) with heaters is connected to the hot leg of the RCS. The full
power hot leg temperature of APR+ was increased from 323.9oC (615oF) of the APR1400 to
326.1oC (619oF) to optimize the RCS design parameters. The total RCS flowrate is increased
to about 103% of the APR1400, which is optimized through the primary component sizing.
In the reactor pressure vessel (RPV) design, four direct vessel injection (DVI) lines are
connected to supply emergency core cooling water from the in-containment refuelling water
storage tank (IRWST). Level probes are added in the hot leg to monitor the water level
during mid-loop operation. The RCS overpressure protection and reactor safety
depressurization functions are established by four (4) pilot-operated safety relief valves
(POSRVs) while the emergency reactor depressurization valves (ERDVs) provide rapid
depressurization function dedicated for severe accidents. On the secondary side of the SGs,
two main steam lines are arranged on each SG, and each steam line has five non-isolable main
steam safety valves (MSSVs), one main steam atmospheric dump valve (MSADV), and two
main steam isolation valves (MSIVs). The APR+ design employs the passive auxiliary
feedwater system (PAFS) that removes core decay heat and RCS residual heat through SGs to
establish and maintain the plant in a safe shutdown condition when the normal feedwater
supply is not available. The PAFS, which is intended to completely replace the conventional
active auxiliary feedwater system, is one of the advanced passive safety features adopted in
APR+. The PAFS cools down the steam generator secondary side and eventually removes the
decay heat from the reactor core by introducing a natural driving force mechanism. A
schematic diagram of arrangements and locations of the primary components and safety-
related systems are shown in Figure 2.

Figure 2. Schematic Diagram of Primary Components and Safety Systems

2.2 Reactor core and fuel design
The reactor core of APR+ is designed to generate 4,290 MW thermal power with an average
volumetric power density of 101.9W/cm3. The reactor core consists of 257 fuel assemblies
made of fuel rods containing uranium dioxide (UO ) fuel. The number of control element
2
assemblies (CEAs) used in the core is 109 in which 97 CEAs are full-strength reactivity
control assemblies and the rest are part-strength CEAs. The absorber materials used for full-
strength control rods are boron carbide (B C) pellets while Inconel alloy 625 is used as the
4
absorber material for the part-strength control rods.
The core is designed for an operating cycle of 18 months or longer with a maximum rod
burnup as high as approximately 60,000 MWD/MTU, and has an increased thermal margin of
more than 10% to enhance safety and operational performance. A portion of the fuel rods
contains uranium fuel mixed with a burnable absorber of gadolinium (Gd O ) to suppress
2 3
excess reactivity after fuelling and to help control the power distribution in the core. The
neutron flux shape is monitored by means of 65 fixed in-core instrumentation (ICI)
assemblies.
The possibility of utilizing the mixed oxide (MOX) fuel up to 1/3 core is considered in the
core design. Eight additional reserve CEAs are installed to increase the reactivity control
capability, if necessary, for MOX fuel loadings. Also, the APR+ reactor core is designed to be
capable of daily load follow operation.
The fuel assembly consists of fuel rods, spacer grids, guide tubes, and upper and lower end
fittings. 236 locations of each fuel assembly are occupied by the fuel rods containing UO
2
pellets or the burnable absorber rods containing Gd O -UO in a 16 × 16 array. The remaining
2 3 2
locations are 4 CEA guide tubes and 1 in-core instrumentation guide tube for monitoring the
neutron flux shape in the core.
The HIPER16TM fuel, which will be used for APR+, has the capability of a batch average discharge
burn-up as high as 65,000 MWD/MTU and has an increased overpower margin as compared to the
previous fuel design (PLUS7TM). The HIPER16TM mid-grid design has high through-grid dynamic
buckling strength for the enhanced seismic performance. The top nozzle has easy reconstitutability
features and hold-down spring force was optimized to reduce the fuel assembly bow. The debris
filtering and capturing features are implemented in the bottom grid by combining the debris filtering
bottom grid and the long bottom end plug to reach the target of zero fuel failure. The bottom nozzle
has a low pressure drop features with rectangular flow holes. The integrity of HIPER16TM fuel has
been enhanced by increasing the fretting wear resistance and debris filtering efficiency. Also, the
safety of HIPER16TM fuel has been enhanced by increasing the seismic performance which is related
to the spacer grid crush strength and dynamic stiffness.
2.3 Fuel handling and transfer systems
The fuel handling system is designed for a safe and rapid handling and storage of fuel
assemblies from the receipt of fresh fuel to the shipment of spent fuel.
The major equipment of the system comprises the refuelling machine, the CEA change
platform, the fuel transfer system, the fresh fuel elevator, the CEA elevator and the spent fuel
handling machine. The refuelling machine is located in the containment building and moves
fuel assemblies into and out of the reactor core and between the core and the fuel transfer
system. The spent fuel handling machine, located in the fuel building, carries fuel to and from
the fuel transfer system, the fresh fuel elevator, the spent fuel storage racks, and the spent fuel
shipping cask areas.
2.4 Primary components
2.4.1 Reactor pressure vessel
The reactor pressure vessel is a vertically mounted cylindrical vessel with a hemispherical
lower head welded to the vessel and a removable hemispherical upper closure head as shown
in Figure 3. The internal surfaces in contact with the reactor coolant are cladded with austenitic
stainless steel to prevent corrosion. The reactor vessel diameter of the APR+ is bigger than that of the
APR1400 to accommodate additional fuel assemblies by about 0.3 m (1 ft). The reactor pressure
vessel contains internal structures, core support structures, fuel assemblies, control rod
assemblies, and control and instrumentation components.
Figure 3. Reactor Pressure Vessel and Internals

The structural integrity of the reactor pressure vessel is verified through a structural sizing and
fatigue evaluation, which calculates the stresses of the heads, shell and nozzles under thermal
and pressure loads.
The direct vessel injection (DVI) nozzles are attached to the reactor vessel for the direct
emergency coolant injection as a part of the safety injection system. The location of DVI
nozzle is above the cold leg nozzles to avoid the interference with reactor vessel external
nozzles and support structure.
The life time of the reactor pressure vessel is extended to 60 years by the use of low carbon
steel, which has lower contents of Cu, Ni, P, S, resulting in an increase of brittle fracture
toughness. The inner surface of the reactor vessel is cladded with austenitic stainless steel or
Ni-Cr-Fe alloy. The reactor vessel is designed to have an end-of-life RT of 21.1oC (70oF).
NDT
The reactor vessel is basically manufactured with a vessel flange, a hemispherical bottom
head, and three shell sections of upper, intermediate and lower. The vessel flange is a forged
ring with a machined ledge on the inside surface to support the core support barrel, which in
turn supports the reactor internals and the core. The three shell sections, the bottom head
forging and vessel flange forging are joined together by welding. Also, four inlet nozzle
forgings, two outlet nozzle forgings, four DVI nozzle forgings, and sixty-one ICI nozzles are
also welded. The upper closure head is fabricated separately and is joined to the reactor vessel
by bolting. The dome and flange are welded together to form the upper closure head, on
which the control element drive mechanism (CEDM) nozzles are welded.
2.4.2 Reactor internals
The reactor internals consist of the core support structures, which include the core support
barrel, upper guide structure barrel assembly and lower support structure, and the internal
structures. The core support structures are designed to support and orient the reactor core fuel
assemblies and control element assemblies, and to direct the reactor coolant to the core. The
primary coolant flows in through the reactor vessel inlet nozzles from the reactor coolant
pump, passes through the annulus between the reactor vessel and core support barrel, through
the reactor vessel bottom plenum and core, and finally flows out through the outlet nozzles of
the reactor vessel connected to the hot legs.
The core support barrel and the upper guide structure are supported at its upper flange from a
ledge in the reactor vessel flange. The flange thickness is increased to sustain the enhanced
seismic requirements. All reactor internals are manufactured of austenitic stainless steel
except for the hold-down ring, which is made of high-tension stainless steel. The hold-down
ring absorbs vibrations caused by the load to the axial direction of internal structures.
The upper guide structure, which consists of the fuel assembly alignment plate, control
element shroud tubes, the upper guide structure base plate, CEA shrouds, and an upper guide
structure support barrel, is removed from the core as a single unit during refuelling by means
of special lifting rig.
2.4.3 Steam generators
The steam generator (SG) is a vertical inverse U-tube heat exchanger with an integral
economizer, which operates with the RCS coolant in the tube side and the secondary coolant
in the shell side as shown in Figure 4. The basic design feature of the APR+ steam generator is the
same as that of the APR1400. The heat transfer area of the SG is increased to accommodate the
uprated thermal power by increasing the height of U-tubes. The height is increased by about one (1)
foot as compared to that of the APR1400. Therefore, the upper shell diameter and the head dome
radius of the SG would increase by 10 inches and 5 inches, respectively. The SG vessel material is
changed from SA-508 Grade 1A to high strength alloy steel, SA-508 Grade 3 class 2, to reduce the
total weight of the SG using the optimized thickness of the SG vessel.
The moisture separators and steam dryers on the shell side of the SG limit the moisture content of the
exit steam to less than 0.25 w/o during normal operation. To preclude uncontrolled steam releases
and resultant uncontrolled cooldown of the RCS, the main steam isolation valves (MSIVs) are
designed as fail-close. Redundant MSIVs to meet the single failure criteria are provided in each main
steam line. The capacity of the spring-loaded main steam safety valves (MSSVs) is sufficient to limit
the maximum steam generator secondary side pressure in accordance with the requirements. The
MSSVs discharge steam to the atmosphere outside the reactor containment building. The leak-
before-break concept is applied to the main steam lines to reduce the need for redundant supports of
the piping and, thus, the construction and maintenance costs will be reduced.
To improve the operability, the angle of nozzle in the hot leg side of the primary system is
modified to enhance the stability of mid-loop operation. The SG water level control system is
designed in such a way that the water level is controlled automatically over the full power
operating range.
The economizer feedwater nozzle provides a passage of feedwater to the economizer, which
is installed to increase the thermal efficiency of the steam generator at the cold side, and
experiences a high temperature gradient. The feedwater nozzles are designed to endure the
excessive thermal stress which causes an excessively large fatigue usage factor. The
downcomer feedwater nozzle attached in the upper shell of SG also provides small portion of
feedwater to the downcomer to facilitate internal recirculation flow. 10% of full power
feedwater flow is provided to the downcomer feedwater nozzle and the remaining feedwater
to the economizer feedwater nozzle at a reactor power higher than 15%, below which all
feedwater is supplied to the donwncomer feedwater nozzle.

Steam Outlet Nozzle
Steam Dryer
Centrifugal Separator Upper Taps, 550.5"
Secondary Manway
Steam Dryer Drain
NWL
Recirculation Nozzle
Downcomer Feddwater Nozzle
0% NR, 400.5"
Downcomer
Tube Support Plate
Evaporator
Tube Bundle Shroud
0% WR, 150.5"
Center Support Cylinder
Economizer
Economizer Feedwater Nozzle
Handhole
Tubesheet Blowdown Nozzle
Nozzle Dam

Figure 4. Steam Generator

2.4.4 Pressurizer
The pressurizer (PZR) is equipped with nozzles for sprays, surgeline, pilot-operated safety relief
valves (POSRVs), emergency reactor depressurization valves (ERDVs), and pressure and level
instrumentations. The RCS overpressure protection and manual safety depressurization functions are
established by four (4) POSRV assemblies while ERDVs provide a rapid depressurization function
dedicated for severe accidents.
The volume of pressurizer (PZR) is 80.7 m3, being larger than the ratio of power increase
from the APR1400, which would reduce the potential causes of plant unavailability by more
effectively accommodating overpressure transients as well as reactor coolant volume
shrinkage or swelling due to temperature/load changes.
2.4.5 Integrated head assembly (IHA)
APR+ adopts the integrated head assembly (IHA) to simplify structural configuration of the upper
closure head region of the reactor vessel and to improve maintenance convenience as shown in
Figure 5. By adopting the IHA, the occupational exposure dose, component storage area and
overhaul duration are expected to be reduced significantly.
The newly developed control element drive mechanism (CEDM) is implemented in the APR+
design. It will be able to operate for a longer period of time and to allow the CEDM motor and coil
assembly to be suitable for the load follow operation. The CEDM power cable is installed at the
upper part of the coil housing. This arrangement can simplify the CEDM coil assembly and enhance
the coil cooling performance. The reduced cooling air flow will also contribute to the simplification
of the RV head area system related to the CEDM because the cooling fan size and seismic loads can
be reduced.

Figure 5. Integral Head Assembly
2.4.6 Reactor coolant pumps
The newly developed RCP is implemented in the APR+ design. This RCP is the same model as, but
larger in size than that that for the Shin-Ulchin Nuclear Power Plant Units 1 and 2 (SUN 1&2), the
first construction plant utilizing new RCPs. It has several improved design features as compared to
the RCP for APR1400 as shown in Figure 6. For example, the use of standstill seal minimizes the
RCS water leakage via the shaft seal under simultaneous loss of seal injection water and component
cooling water after and/or during the station blackout. The spool piece, which connects the pump
shaft to the thrust bearing shaft, provides enough space to replace the mechanical seal without lifting
the thrust bearing assembly.
To find an optimal hydraulic design that satisfies all of the hydraulic performance requirements of
APR+ RCP, an RCP model test has been performed. The APR+ RCP model test data are the
reference for guaranteeing performance of prototype RCP. Another purpose of the model test is to
acquire the 4-quadrant curve of the accepted hydraulic model. The 4-quardrant curve is the
performance curve for all the possible pump flow and rotation conditions. The 4-quadrant curve is
used for input data for safety analysis.
Figure 6. SKN 3&4 RCP(left) vs. APR+ RCP(right)

2.4.7 RCS Piping
Since the pipe whip restraint and the support of the jet impingement shield in the piping
system of earlier plants are expensive to build and maintain and could lead to a potential
degradation of plant safety, the Leak-Before-Break (LBB) principle is adopted for the piping
system of APR+. The LBB principle is applied to the main coolant lines, surge lines, and
pipes in the shutdown cooling system and the safety injection system. The application of
LBB reduces the redundant supports of the pipe in the NSSS pipe system since the dynamic
effects of postulated ruptures in the piping system can be eliminated from the design basis.
Therefore, the cost of design, construction and maintenance is reduced.
2.5 Reactor auxiliary systems
2.5.1 Chemical and volume control system (CVCS)
The CVCS of the APR+ is not required to perform any safety functions such as safe shutdown and
accident mitigation. This system is basically for the normal day-to-day operation of the plant. The
components related to charging and letdown functions, however, are designed as a safety grade and
reinforced to assure the reliability for normal and transient conditions. Two centrifugal charging
pumps and a flow control valve provide required charging flow. For normal operation, only one
charging pump is used to supply the required minimum flow of 12.6 kg/s.
Pressure reduction of the letdown flow from the reactor coolant system occurs at the letdown orifice
and the letdown control valve. Then the letdown flow passes through the regenerative and letdown
heat exchangers, where the temperature reduction takes place. Following pressure and temperature
reduction, the flow passes through a purification process at the filters and ion exchangers. After
passing through the purification process, the letdown flow is diverted into the volume control tank
(VCT) which is designed to provide a reservoir of reactor coolant for the charging pumps and for the
dedicated seal injection pumps for the reactor coolant pumps.
2.5.2 Component cooling water system